The fuel assemblies of water-cooled nuclear reactors, such as pressurized-water nuclear reactors, comprise a framework in which fuel rod of great length are disposed in order to form a bundle. The framework comprises spacer grids which are spaced relative to one another along the length of the assembly and connected together by guide tubes.
Each of the spacer grids comprises an assembly of cells each intended to receive a fuel rod and disposed in a uniform network, generally with squared mesh.
At the level of each of the cells intended to receive a fuel rod, the spacer grids comprise means for gripping the pencil, while also ensuring both transverse retention and longitudinal retention of this rod. These gripping means generally consist of bosses projecting inwards relative to the walls of the cell of the grid and of springs consisting of resilient elements formed in the metal of certain walls of the cells of the grid or, alternatively, connected to these walls.
When a new fuel assembly is charged in the nuclear reactor, it has mechanical and physical characteristics which are in accordance with the manufacturing and design standards which are defined or calculated so that the assembly can fulfill its function under the operating conditions of the reactor without suffering any excessive damage during a specific number of operating cycles.
Bearing in mind the irradiation of the materials, the mechanical characteristics of the constituents of the reactor evolve but remain within specific limits ensuring safety of operation.
In particular, with regard to the fuel assemblies, the spring and boss assemblies distributed in a uniform manner in the cells of the grids must retain mechanical characteristics and, in particular, strength and resilience characteristics which are sufficient to ensure effective retention of the fuel rods during a long operating period of the reactor.
However, in order to improve safety and performance in nuclear reactors, it has been considered that it was desirable to check, periodically, the efficiency of the means for gripping the fuel rods in the assemblies in the core of the reactor.
This checking must be performed during a discharging and refueling operation of the nuclear reactor, the assemblies being placed in a storage pool under a certain depth of water which makes it possible to ensure the biological protection of the operators responsible for the checking.
A purely visual examination performed, for example, using video means, does not make it possible to detect those fuel rods whose gripping has become insufficient and, in particular, those rods on which is exerted an axial retention force which no longer makes it possible to ensure effective retention of the rod during operation of the reactor. In fact, in this case, the rod may be caused to vibrate by the passage of the cooling fluid of the reactor, such that the rod which has been caused to vibrate is liable to suffer breakage. In this case, the pellets of uranium disposed in the sheath of the rod are dispersed in the primary shell of the nuclear reactor and cause high levels of contamination in this primary shell.
A device described in U.S. Pat. No. 4,265,010 is known, which makes it possible to replace, in correct position, those fuel rods which have undergone a certain axial slide. This device comprises two parallel plates, the gap between which may be varied by means of a displacement control device. One of the plates rests on an end joining piece of the assembly and the other plate, comprising a centering dish, rests under the end of the rod whose upward displacement inside the assembly is to be ensured in order to replace it in correct position. However, such a device does not make it possible to measure the gripping and longitudinal retention force on the rod in order to determine whether there is a risk of displacement of and damage to the rod during refueling of the assembly in the nuclear reactor.
Patent EP-A-0,l46,804 describes a remotely controlled device for lifting fuel rods. The fuel rods are displaced so as to cause the support points of the sheath to pass over the gripping means above the grid of the assembly in order to inspect possible defects in the sheath by video means. This device does not make it possible directly to measure the axial retention force on the fuel rod.
Moreover, the device, which is awkward to use, can be used only at the level of the lower joining piece of the assembly and does not make it possible to lower the rods back into their initial position.
Moreover, it is desirable to be able to perform checks on the assemblies without having to remove their end joining pieces; this operation is possible only in respect of specially designed assemblies and, in any case, requires a complicated work tool and procedure. It is thus necessary to limit checking to the peripheral rods of the assembly which are the most stressed and the most likely to have gripping defects after a certain residence time in the operating reactor.
In particular, devices are known which comprise a means for axial pushing on an end of a rod and means for measuring the pushing force on the rod which may be operated and activated remotely. However, such devices are not designed to perform precise measurements.